Due to a large plutonium (Pu) throughput and high burn-up fuel in an advanced reprocessing plant, we have the responsibility to undertake the inevitable burden of nuclear material accountancy (NMA) to meet the International Atomic Energy Agency (IAEA) safeguards criteria. A large amount of sampling analysis and inspectors' activities result in a great cost in facility operation to verify no concealment and undeclared use of Pu. In addition to NMA, containment and surveillance (C/S), process monitoring (PM), and curium (Cm) balance have been used for the safeguards activities to complement NMA. However, except for NMA, any mathematical or regulatory formalism in the safeguards measures have not been presented so far, therefore it is difficult to evaluate the cost-effective performance of the safeguards system. In order to design an advanced safeguards system for the fast reactor fuel cycle, the JAEA has started to develop a safeguards system simulator. A NMA core in the simulator is composed of a near-real-time accounting (NRTA) code which had been already developed and applied to investigate the NMA characteristics of JAEA facilities. A "multivariate and multi-scale core" is based on a multivariate mathematical analysis combined with a multi-scale statistical process analysis making use of a wavelet decomposition forms safeguards envelope, which provides a control and monitoring system logic. Multi-scale principle component analysis of the core had been applied to "material unaccounted for" (MUF). A concept of multiple optimization core is proposed as the safeguards formalism, with probabilistic risk analysis and cost-performance characteristics of the safeguards system, is discussed in the presentation. Flow meter and non-destructive analysis can be more broadly applied to the system in a cost-effective manner. A virtual design and objective-driven model will be developed in the simulator in the future to support an effective safeguards design and to develop a "walk-through" virtual plant model.

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The evaluation of uncertainty constitutes the necessary supplement of Best Estimate (BE) calculations performed to understand accident scenarios in water cooled nuclear reactors. The needs come from the imperfection of computational tools on the one side and from the interest in using such tool to get more precise evaluation of safety margins. In the present paper the approaches to uncertainty are outlined and the CIAU (Code with capability of Internal Assessment of Uncertainty) method proposed by the University of Pisa is described including ideas at the basis and results from applications. Two approaches are distinguished that are characterized as "propagation of code input uncertainty" and "propagation of code output errors". For both methods, the thermal-hydraulic code is at the centre of the process of uncertainty evaluation: in the former case the code itself is adopted to compute the error bands and to propagate the input errors, in the latter case the errors in code application to relevant measurements are used to derive the error bands. The CIAU method exploits the idea of the "status approach" for identifying the thermal-hydraulic conditions of an accident in any Nuclear Power Plant (NPP). Errors in predicting such status are derived from the comparison between predicted and measured quantities and, in the stage of the application of the method, are used to compute the uncertainty.

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Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers and vendors, nuclear fuel companies, research organizations, consulting companies, and technical support organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the 'user effect' and stems from the limitations embedded in the codes as well as from the limited capability of the analysts to use the codes. Code user training and qualification is an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. The 3D S.UN.COP (Scaling, Uncertainty and 3D COuPled code calculations) seminars have been organized as follow-up of the proposal to IAEA for the Permanent Training Course for System Code Users. Six seminars have been held at University of Pisa (2003, 2004), at The Pennsylvania State University (2004), at University of Zagreb (2005), at the School of Industrial Engineering of Barcelona (January-February 2006) and in Buenos Aires, Argentina (October 2006), being this last one requested by ARN (Autoridad Regulatoria Nuclear), NA-SA (Nucleoelectrica Argentina S.A) and CNEA (Comision Nacional de Energia Atomica). It was recognized that such courses represented both a source of continuing education for current code users and a mean for current code users to enter the formal training structure of a proposed 'permanent' stepwise approach to user training. The 3D S.UN.COP 2006 in Barcelona was successfully held with the attendance of 33 participants coming from 18 countries and 28 different institutions (universities, vendors, national laboratories and regulatory bodies). More than 30 scientists (coming from 13 countries and 23 different institutions) were involved in the organization of the seminar, presenting theoretical aspects of the proposed methodologies and holding the training and the final examination. A certificate (LA Code User grade) was released to participants that successfully solved the assigned problems. A seventh seminar is currently holding (January-February 2007) at the Texas A&M University involving more than 30 scientists between lecturers and code developers. (http://dimnp.ing.unipi.it/3dsuncop/2007)

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The economic advantages of a chordal ultrasonic feedwater flow measurement system over conventional (flow nozzle-based) feedwater instrumentation are analyzed for new plants having ratings ranging from 1100 MWe to 1600 MWe. Specifically, each of the following topics is considered: ・The value of a 1.7% increase in the rating of the new plant, made possible by the reduced uncertainty in the determination of thermal power. ・The value of reduced startup time owing to enhanced steam supply water level control. ・The value of the reduced feedwater pumping power brought about by the elimination of flow nozzles. ・The value of the reduced calibration burden owing to the elimination of the feedwater flow differential pressure transmitters and resistance thermometers ・The net difference in the acquisition costs of the ultrasonic system versus conventional feedwater flow instrumentation. ・The net savings in installation costs of the ultrasonic system vis-a-vis conventional feedwater flow instrumentation. ・The potential savings in outage time due to the reduced frequency of low steam supply water level trips (scrams) of the reactor.

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Recently the JSME has published the guideline for the high-cycle thermal fatigue in pipelines in nuclear power plants, which is caused by the temperature fluctuation in mixing tees with high difference of temperature between two fluids. Many experimental studies have been conducted to establish the evaluation method for the thermal fatigue of pipe. The cavity flow is one of typical flow patterns which cause the high-cycle thermal fatigue. It is driven by the main flow as a stagnant branch flow, where the other end of the branch pipe connected to the main pipe is closed by a valve and so on. When the flow in the main pipe is hotter than the branch flow, heat of the main flow is transported to the branch flow. According to experiments, the thermal stratification is formed in the branch pipe and the stratification surface fluctuates. These phenomena will threaten the structural integrity of the piping system. In order to estimate the high-cycle thermal fatigue of the pipe wall, it is necessary to predict the location of thermal stratification, the amplitude of temperature fluctuation and its frequency. The purpose of this study is to clarify the mechanism of this phenomena using numerical fluid dynamics simulation and to predict the location of thermal stratification. We evaluated the effect of various numerical turbulent models on the branch flow and found that proper numerical modeling of shear force by the main flow and the velocity distribution on the wall boundary in the branch pipe are important for high-accuracy prediction of the location of the thermal stratification. The paper will discuss the numerical modeling and the results compared with experimental results for the cavity flow.

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Measurements of two-phase flow parameters such as void-fraction, bubble velocities, and interfacial area densities have been performed in an upwards air-water flow at atmospheric pressure by means of a four-tip needle-probe and a wire-mesh sensor. For the first time, a direct comparison between the two measuring techniques has been carried out. Both techniques are based on the measurement of the fluid conductivity. For void fraction and velocity measurements, similarity exists between the two methodologies for signal analysis. A significantly different approach is followed, instead, for the estimation of the interfacial area concentration: while the evaluation based on the needle-probe signal is carried out by using projections of the gas-liquid interface velocity, the evaluation based on the wire-mesh signals consist in a full reconstruction of the bubbles interfaces. The comparison between the two techniques shows a good agreement.

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In order to derive validation data for numerical code validation the REKO-3 facility at Forschungszentrum Juelich (FZJ) is used for experimental studies on recombiner behaviour under well defined steady-state conditions. One important aspect addressed in recent experiments is the influence of an oxygen depletion on the catalytic reaction. Detailed measurements of the catalyst temperature and the hydrogen concentration along the catalyst sheets have been performed for oxygen concentrations from oxygen content in air down to 1 vol.% at different flow rates. The critical oxygen concentration that marks a loss in performance of the recombiner has been determined in dependency on several influence parameters. Due to the local measurements the influence of oxygen depletion on the reaction kinetics is already detected before the integral reaction rate is affected. The numerical model REKO-DIREKT has been used for recalculation of the experiments. The code was capable to accurately calculate the influence of the oxygen concentration on the catalytic reaction by achieving good agreement with the experimental data.

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Containments of German light water reactors (LWR) are equipped with passive auto-catalytic recombiners (PAR) for the removal of hydrogen released during a severe accident. The exothermal reaction of hydrogen and oxygen on the catalyst sheets inside the PAR provides a mitigation of the threat of a detonation. A research project 'Studies for the advancement of the safety assessment methods for PAR in NPP containments during severe accidents' is presently conducted at the RWTH Aachen University, Chair for Reactor Safety and Reactor Technology (LRST), in collaboration with the Institute for Safety Research and Reactor Technology (ISR) at the Forschungszentrum Juelich GmbH (FZJ) and funded by the German Federal Ministry of Economics and Technology (BMWi). The project aims at the achievement of a profound understanding of recombiner phenomena in order to develop advanced models for the PAR assessment in severe accident analysis. The work programme includes several experimental studies at different test facilities used in cooperation with FZJ. Aspects to be considered are e.g. the starting behaviour, the inerting effect of steam and the effect of different geometries of the catalyst elements. A medium scale facility for studies under natural flow conditions is currently under construction. The experimental results obtained will be used for code development and validation.

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To evaluate the effect of surface wettability in pipe wall on flow characteristics in a vertical upward gas-liquid two-phase flow, visualization study was performed using three test pipes, namely an acrylic pipe, a hydrophilic pipe, a hydrophobic pipe. Such basic flow characteristics as flow patterns and void fraction were investigated in these three pipes. In the hydrophilic pipe, the slug flow-to-churn flow transition boundary was shifted to higher gas velocity condition at a given liquid velocity, whereas the churn flow-to-annular flow transition boundary was shifted to lower gas velocity condition at a given liquid velocity. In the hydrophobic pipe, the inverted-churn flow regime was observed in the region where the churn flow regime was observed in the acrylic pipe, whereas the droplet flow regime was observed in the region where the annular flow regime was observed in the acrylic pipe. At high-gas flow rate condition, the mean void fraction in the hydrophobic pipe took relatively higher value to that in the acrylic pipe.

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In our previous study, significant improvements of surface wettability and boiling heat transfer on oxide film coated-materials by Radiation Induced Surface Activation (RISA) phenomenon were confirmed in a room temperature condition. To delineate the RISA effect in actual boiling systems or at high temperatures and pressures, this study aimed at obtaining contact angle, an indicator of macroscopic surface wettability, of water droplet on aluminum, zircaloy and stainless plates before and after γ-ray irradiation at ambient temperatures up to 280 ℃. The contact angle was measured by an image-processing of the images obtained by a CCD video camera. In condition less than 200 ℃, super-hydrophilicity or null contact angle was achieved after 830 kGy γ-ray irradiation. On the other hand, at over 200 ℃, super-hydrophilicity by the RISA effect was not achieved even after 2800 kGy γ-ray irradiation, although slight improvement of surface wettability was confirmed.

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Process Data Reconciliation (PDR) is a certified method that calculates the most likely values considering process measurement uncertainties and continuously closing all energy-, mass- and material balances. All interdependencies between measurements within the entire plant process are mapped in a covariance matrix. There are several areas where monetary and non-monetary benefits for the plant operator are generated: 1. Safety 2. Transparency 3. Financial benefits Transparency The "true" state of the process is becoming visible. The traditional belief in absolute measurement values is changed into a realistic view on values that are addressed with uncertainties. With this, new insights are won in process areas, which are not or insufficiently instrumented. Process data reconciliation deduces missing process data by regarding the entire process, all its redundancies and their relationship to each other. This creates a detailed understanding for the processes and interactions within the processes. Safety The methodology of process data reconciliation represents the latest development level. It is the most precise tool for process control & analysis and is used world-wide. VALI and ProcessPlus^<TM> are certified according to VDI2048 and tested by several national institutions. The more transparent process data becomes to the operator, the safer he can control the plant. Online process data reconciliation enables the view on true process data and gives instant indications to sub-optimal process behaviour as well as malfunctioning components and leakages. Furthermore, the ability to instantly differentiate between a process change and a drifting measurement is crucial to the safe operation of nuclear power plants. Financial benefits Financial benefits are created by several aspects: Plant efficiency is increased by the rise in process transparency. Suboptimal conditions in the plant are rapidly located. Minimizing uncertainty of measurement values makes it safer to approximate limit values more closely thus enabling the operator to maximize the output. Furthermore, there is a substantial reduction in maintenance cost enabled through process transparency. The shift from preventive to conditionbased maintenance can be safely executed. In addition, contracts with suppliers concerning retrofit measures need to be based on the VDI 2048 guideline to forgo customer penalizing. Highly accurate and low-cost acceptance tests with process data reconciliation are performed using operational data only and ensure an independent view on the conducted measures. Process data reconciliation is per Gauss' definition the best possible tool to describe the true state of a plant's process. Highly precise measurements can only contribute to sharpening the total view, but never exceed the overall precision of data reconciliation.

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The experiment LACE LA4 on thermal-hydraulics and aerosol behavior in a nuclear power plant containment, which was performed in the LACE experimental facility, was simulated with the ASTEC CPA module of the severe accident computer code ASTEC V1.2. The specific purpose of the work was to assess the capability of the module (code) to simulate thermal-hydraulic conditions and aerosol behavior in the containment of a light-water-reactor nuclear power plant at severe accident conditions. The test was simulated with boundary conditions, described in the experiment report. Results of thermal-hydraulic conditions in the test vessel, as well as dry aerosol concentrations in the test vessel atmosphere, are compared to experimental results and analyzed.

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In a sodium-cooled fast reactor, highly sensitive technology is required to detect a sodium leak from the cooling system piping or components. Conventional sodium leak detectors have difficulty in measuring small amounts of sodium leak because of the presence of salinity in the atmosphere. In order to overcome this problem, an innovative technology has been developed to selectively detect the radioactive sodium of the primary cooling system using Laser Resonance Ionization Mass Spectrometry (RIMS). This research and development program consists of investigating the detection process of sodium aerosol by RIMS, manufacturing the prototype sodium detection system, and testing its function using an actual radioactive sodium sample, which will be taken from the experimental fast reactor Joyo. The schedule and the results obtained from the investigation of the sodium leakage detection process are shown in this paper.

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Mechanical loadings on pipe systems caused by water hammer with phase changes make calculation of final forces difficult in nuclear power plants. The common procedure in Sweden is to calculate the water hammer loadings, according to the classical one-dimensional theory of liquid transient flow in a pipeline, and then transfer the results to strength analyses of pipeline structure. This procedure assumes that there is quasi-steady response of the pipeline structure to pressure surges - no dynamic interaction between the fluid and the pipeline construction. The hydraulic loadings are calculated with 1-D so-called "network" programs. Commonly used in Sweden are Relap5 (Mod3.2.2 and higher) and Drako. As a third party accredited inspection body INSPECTA NUCLEAR AB reviews calculations of water hammer loadings. An important question for the reviewer (and also for the users) is knowledge about their ability to calculate the dynamic loadings. While the ability of Relap5 and DRAKO to calculate water hammer without phase changes is relatively well investigated the skills of the programs when phase changes are present need some more attention. The presented work shall be seen as an attempt to illustrate ability of Relap5, and Drako programs to calculate the water hammer loadings with phase changes. A special attention was paid to using of Relap5 for calculation of water hammer pressure surges (including some aspects of influence of discretisation of space on the calculation results). The calculations are compared with experimental results. The experiments have been conducted at a test rig designed and constructed at the Szewalski Institute of Fluid-Flow Machinery of the Polish Academy of Sciences (IMP PAN) in Gdansk, Poland. The comparison of calculated and measured pressures shows some differences, only the first pressure peak, occurring before evaporation is calculated quite exactly. All next coming pressure peaks differ slightly from the measured with respect to amplitude and frequency.

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The paper describes the modelling features for a VVER-1000 nuclear power plant by the coupled system code ATHLET/BIPR-VVER. Special attention is paid on the reactor pressure vessel model and its nodalization schema. They play an important role for the realistic prediction of the fluid mixing phenomena in case of asymmetric transients. For such a transient - simultaneous isolation of one steam generator from the steam line and the feed water supply at low power level - validation to measured parameters are presented and analysed. The results of the simulation are compared with experimental local and integral plant loop data from NPP Kozloduy Unit 6 collected during the plant-commissioning phase. Additional examples of coupled code application for VVER-1000 reactor applying detailed reactor pressure vessel nodalization are also presented, choosing Exercise 3 of the CEA-NEA/OECD VVER-1000 Coolant Transient Benchmark (V1000CT-2) - a main steam line break.

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TOSQAN is an experimental program undertaken by the Institut de Radioprotection et de Surete Nucleaire (IRSN) in order to perform thermal hydraulic containment studies. The TOSQAN facility is a large enclosure devoted to simulating typical accidental thermal hydraulic flow conditions in nuclear Pressurized Water Reactor (PWR) containment. The TOSQAN facility, which is highly instrumented with non-intrusive optical diagnostics, is particularly adapted to nuclear safety CFD code validation (Porcheron et al., 2003). The present work is devoted to studying the interaction of a water spray injection used as a mitigation means in order to reduce the gas pressure and temperature in the containment, to produce gases mixing and washout of fission products. In order to have a better understanding of heat and mass transfers between spray droplets and the gas mixture, and to analyze mixing effects due to spray activation, we performed detailed characterization of the two-phase flow.

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In the framework of the HTTR (High Temperature Engineering Test Reactor) project, Japan Atomic Energy Agency has been conducting R&D on thermo-chemical Iodine-Sulfur (IS) process for large scale hydrogen production. Since the IS process constitutes corrosive environments, screening of the materials of construction is very important. Corrosion resistance of lining materials, were examined in 47, 75, 90wt% sulfuric acid for 5 hours at temperatures ranging from 200℃ to 400℃ under inert gas pressure of 2 MPaG. Gold plating materials showed good corrosion resistance irrespective of the base metal, Alloy B2 and SUS304. Soda glass for glass lining material was corroded in 47wt% sulfuric acid at 200℃. However, it showed good corrosion resistance in 75wt% sulfuric acid at 300℃ and in 90wt% sulfuric acid at 400℃.

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A thermal-hydraulic feasibility project for an Innovative Water Reactor for Flexible fuel cycle (FLWR) has been performed since 2002. In this R&D project, large-scale thermal-hydraulic tests, several model experiments and development of advanced numerical analysis codes have been carried out. In this paper, we will describe the critical power characteristics in a 37-rod tight-lattice bundle with rod-bowing under both steady and transient states. It is observed that no matter it is run under a steady or a transient state, boiling transition (BT) always occurs axially at exit elevation of upper high-heat-flux region and transversely in the central area of the bundle. Steady critical power increases monotonically with the increase of mass velocity, with the decrease of inlet water temperature and with the decrease of exit pressure. These trends are same as those in the base case test without rod-bowing. The steady critical power with rod-bowing is about 10% lower than that without rod-bowing. For the postulated power increase and flow decrease cases that may be possibly met in a normal operation of the FLWR, it is confirmed that no BT occurs when Initial Critical Power Ratio (ICPR) is 1.3. Moreover, when the transients are run under severer ICPR that causes BT, the transient critical powers are generally same as the steady ones. The experiments are analyzed with TRAC-BF1 code. The TRAC-BF1 code shows good prediction for the occurrence or the non occurrence of the BT and predicts the BT starting time within the accuracy of critical power correlation. Traditional quasi - steady state prediction of the transient BT is confirmed being applicable for the postulated abnormal transient processes in the tight lattice bundle with rod - bowing.

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In order to study the interactions between a spray and an atmosphere representative of a severe accident in a Pressurized Water Reactor, in terms of pressure, temperature and composition (steam and aerosol), the French Institute for Radiological Protection and Nuclear Safety (IRSN) developed the TOSQAN facility. This paper presents the development and qualification of the global rainbow refractometry and Interferometric Laser Imaging for Droplets Sizing (ILIDS) that are respectively dedicated to measure the spray droplets temperature and size. In addition we present an extension of these two techniques in order to determine the aerosol concentration inside the droplet and the aerosol removal rate.

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The real toughness response of RPV material can only be determined after the final shut down of the NPP. Such a chance is given now through the investigation of material from the former Greifswald NPP (VVER-440/230). In the first part the paper deals with the retrospective dosimetry based on Niobium, which is a trace element of the RPV material. Unfortunately, a second neutron reaction besides ^<93>Nb(n,n') leading to ^<93m>Nb-activity is the reaction ^<92>Mo(n,γ)^<93>Mo. Therefore, the ^<93m>Nb-activity generated through the Mo path has to be determined separately. Based on the found Nb and Mo contents, it turns out that the ^<93m>Nb generation on the Mo path dominates in the most cases over the fast neutron induced generation from Nb. The comparison between the calculated and the measured ^<93m>Nb activities typically resulted in deviations of 50%. Possible reasons for the observed differences are discussed. In the second part the first results of fracture mechanic investigations are reported. In a first step SE(B) specimens from three thickness positions were tested and evaluated according to the test standard ASTM E1921-05. Cleavage fracture toughness values, K_<Jc>, were determined and Master Curve based reference temperatures (T_0) were evaluated. The T_0 measured on the inner surface of this RPV does not show the highest value and, thus, reflect the conservative condition. The T_0 of disc B located between the surface and 1/4 thickness is about 40K higher compared with the surface. The K_<Jc> values adjusted to a specimen thickness of 1T are enveloped by the WWER specific lower bound fracture toughness curve suggested in the VERLIFE procedure. Finally can be concluded, that for a successful application of Nb dosimetry a minimum Nb content of 20 ppm is needed under the given conditions (i.e. Mo content and measuring date). The measured K_<Jc> values are not enveloped by the 5% fractile indexed with T_0 + 2σ according to the Master Curve concept. However, the 5% fractile indexed with the VERLIFE reference temperature RT_<To> that includes an additional margin envelops the measured K_<Jc> values. Therefore the VERLIFE lower bound curve conservatively describes the fracture toughness of the investigated weld metal.

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The main objective of this work is to identify and analyze two-phase flow regimes in adiabatic upward flow conditions in an concentric annulus geometry with an inner diameter of 19.1 mm and an outer diameter of 38.1 mm, at three axial locations (z/D_H: 51.5, 148.8, 229.9). In this regard, both Global Flow Regimes (GFRs) and Local Flow Regimes (LFRs) have been identified. In both cases, neural networks techniques have been used to obtain objective identification results. However, the flow regime indicator will be different depending on the type of flow regime: the cumulative probability distribution function (CPDF) of the area averaged void fraction obtained by impedance meters in the case of GFRs and the CPDF of the bubble chord length measured by a four-sensor conductivity probe in the case of LFRs. The GFRs maps have been compared with those obtained in previous research works and flow regime transitions existing correlations. The LFRs have been identified in five radial locations. The LFRs combinations maps have been compared with the GFRs and those combinations obtained in round pipes.

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This paper presents AREVA NP's 30 year experience in the field of decontamination. This paper is focused on decontamination prior to decommissioning and the highlights of the performed projects and results will be outlined. Advantages of the FSD short after final shutdown will be described. Since 1986 AREVA NP has been working regularly on the decontaminations prior to decommissioning and our decommissioning experience in this field cover all main NPP designs: ・1 application in a heavy water PWR of Siemens Design ・1 application in PWR of VVER design ・3 applications in PWR of Siemens design ・3 applications in PWR of Westinghouse design ・5 applications in BWR of GE design It will be demonstrated that decontaminations can be performed with CORD Family and AMDA even after 20 years of safe enclosure (see FSD Lingen and MZFR). Decontaminations can be performed either by using NPP systems / components or completely by using external decon equipment AMDA. In this context, CORD[○!R] (Chemical Oxidation Reduction Decontamination) represents the chemical decontamination process while AMDA[○!R] stands for Automated Mobile Decontamination Appliance. HP is used for permanganic acid as an oxidizing agent and UV for the in-situ decomposition of the decontamination chemicals with ultraviolet light.

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The JT-60SA aims to contribute and supplement ITER toward DEMO reactor based on tokamak concept. One of the features of JT-60SA is its high power long pulse heating, causing the large annual neutron fluence. Because the expected dose rate at the vacuum vessel (VV) may exceed 1 mSv/hr after 10 years operation and three month cooling, the human access inside the VV is prohibited. Therefore a remote handling (RH) system is necessary for the maintenance and repair of in-vessel components. This paper described the RH system of JT-60SA, especially the expansion of the RH rail and exchange of the divertor modules. The RH rail is divided into nine and three-point mounting. The nine sections can cover 225 degrees in toroidal direction. A divertor module, which is 10 degrees wide in toroidal direction and weighs 500kg itself due to the limitations of port width and handling weight, can be exchanged by heavy weight manipulator (HWM). The HWM brings the divertor module to the front of the other RH port, which is used for supporting the rail and/or carrying in and out equipments. Then another RH device receives and brings out the module by a pallet installed from outside the VV.

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The radiation generated by EAST tokamak in 2006 autumn engineering commissioning included photons generated by bremsstrahlung and a small amount of neutrons induced by photon. For appraising the personal doses of the operators and the environmental radiation exposure, LiF (Mg, Cu, P) thermoluminescence detectors (TLDs) were used to measure the dosage caused by photons in relevant areas, including the EAST tokamak hall, the working area (central control room) and the surrounding environment. The measurement results showed that the dose in the tokamak hall was high and the radiation might cause hazard to the health of the people during the discharge period, so any one who entries into or stays in the tokamak hall must be prohibited automatically by the safety interlock system. The measuring results also indicated that the exposures in the central control room and the environment were almost the same as the background radiation, the possible hazards of radiation to the staff and the surrounding inhabitants were very small.

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On WWER type reactors reactor pressure vessel (RPV) integrity should be assured during operational life-time since there is no any real measures which could mitigate consequences of RPV rupture. One of the possible ways of RPV rupture is cold overpressure phenomena. Evaluation of loads/impact on RPV depends on plant condition and initiating events which could lead to cold overpressure. The main characterized parameters of cold overpressure phenomena are high pressure and low temperature. Selection of initiating events which could lead to cold overpressure during full power operation and hot shutdown and identification of accidental scenarios using probabilistic methods are presented in this paper.

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Increases in the power generation efficiency of nuclear power plants (NPPs) are mainly limited by the permissible temperatures in nuclear reactors and the corresponding temperatures and pressures of the coolants in reactors. Coolant parameters are limited by the corrosion rates of materials and nuclear-reactor safety constraints. The advanced construction materials for the next generation of CANDU reactors, which employ supercritical water (SCW) as a coolant and heat carrier, permit the improved "steam" parameters (outlet temperatures up to 625℃ and pressures of about 25 MPa). An increase in the temperature of steam allows it to be utilized in thermochemical water splitting cycles to produce hydrogen. These methods are considered by many to be among the most efficient ways to produce hydrogen from water and to have advantages over traditional low-temperature water electrolysis. However, even lower temperature water splitting cycles (Cu-Cl, UT-3, etc.) require an intensive heat supply at temperatures higher than 550-600℃. A sufficient increase in the heat transfer from the nuclear reactor to a thermochemical water splitting cycle, without jeopardizing nuclear reactor safety, might be effectively achieved by application of a heat pump, which increases the temperature of the heat supplied by virtue of a cyclic process driven by mechanical or electrical work. Here, a high-temperature chemical heat pump, which employs the reversible catalytic methane conversion reaction, is proposed. The reaction shift from exothermic to endothermic and back is achieved by a change of the steam concentration in the reaction mixture. This heat pump, coupled with the second steam cycle of a SCW nuclear power generation plant on one side and a thermochemical water splitting cycle on the other, increases the temperature of the "nuclear" heat and, consequently, the intensity of heat transfer into the water splitting cycle. A comparative preliminary thermodynamic analysis is conducted of the combined system comprising a SCW nuclear power generation plant and a chemical heat pump, which provides high-temperature heat to a thermochemical water splitting cycle for hydrogen production. It is concluded that the proposed chemical heat pump permits the utilization efficiency of nuclear energy to be improved by at least by 2% without jeopardizing nuclear reactor safety. Based on this analysis, further research appears to be merited on the proposed advanced design of a nuclear power generation plant combined with a chemical heat pump, and implementation in appropriate applications seems worthwhile.

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The paper deals with a new approach to modeling the heat transfer phenomena by means of differential equations with delays. The infinite order dynamics of thermal processes by suitable combinations of capacitance and delay elements is presented. An identification of transfer function of heat exchangers is presented. In the mathematical treatment of heat transfer systems, it is usually quite advantageous to deal in the frequency domain rather than the time. In such cases, the response of the system to sinusoidal inputs over a band of frequencies must be known. Identification is based on the least square method, which is based on minimization of the weighted sum of the squares of the errors between the absolute magnitudes of the frequency characteristic real object and the frequency characteristic of time - delay model of heat transfer system, which is proposed in this paper.

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A dynamic thermal-hydraulic mathematical model of evaporator dynamics of a once - through sub critical steam generator is derived and presented. This model allows the investigation of evaporator dynamics including its transients responses. The evaporator was considered as a part of three-section (economizer, evaporator and super-heater) model with time varying phase boundaries and is described by a set of linearized discrete - difference equations which, with some other algebraic equations, constitutes a closed system of equations possible for exact computer solution. This model has been derived upon the fundamental equations of mass, energy and momentum balance. For the first time, a discrete differential approach has been applied in order to investigate such complex, two phase processes. Namely, this approach allows one to escape from the model of this process usually described by a set of partial differential equations and enables one, using this method, to simulate evaporators dynamics in an extraordinarily simple way. In current literature this approach is sometimes called physical discretization.

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The ability of microwave, to propagate inside the pipe above cutoff frequency, makes microwave nondestructive testing (NDT) techniques very suitable to detect the crack in the straight pipes. The detection of open semi circumferential crack and prediction of related locations were confirmed when microwave with circular TM_<01>-mode was generated in the system. In this study circular TE_<11>-mode, which is also generated and can be resonated in the mode-converter, is used for detection of longitudinal crack in straight pipe. The goal of this investigation is to demonstrate the capability of this method to detect longitudinal crack in piping system. More emphasis is put on the evaluation of crack location by measuring time of flight (TOF) of electromagnetic wave inside the inspected pipe. The results for two crack location at several plunger locations either in frequency domain or time domain to measure TOF of the wave are presented. The experimental results of TOF are compared with calculation ones to show the applicability of this technique. The results show good agreement between experimental and calculations.

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Film boiling heat transfer coefficients are obtained and the vapor film behavior is visualized through a series of quenching experiments. The test section is of a downward heated curved surface of which the curvature radius and edge angle are 990 mm and 9.88°, respectively. The copper having the Biot number less than 0.1 in the film boiling regime, heat transfer coefficients are determined without considering conduction in the test section. A total of seven thermocouples are installed near the outer surface of the test section. The test results are compared with those from the laminar film boiling analysis and a previous experiment with the same edge angle. Higher heat transfer coefficients than are predicted by laminar film boiling analysis are measured since the concentric interfacial waves limit the increase in the vapor film thickness by way of an intermittent breakup allowing the heated surface to be rewetted.

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The feed water mass flow is the key measured variable used to determine the thermal reactor output in a nuclear power plant. Usually this parameter is recorded via venturi nozzles or orifice plates. The problem with both principles of measurement, however, is that an accuracy of below 1 % cannot be reached. In the case of nuclear power plants and depending on the size of the plant, this corresponds to an electrical output of 4 MWel to 16 MWel. In order to make more accurate statements about the feed water amounts recirculated in the water-steam circuit, tracer measurements that offer an accuracy of up to 0.2 % are used. A drawback of this method is that this measuring principle is suitable only for providing an instantaneous picture but does not provide continuous operating information about the feed water mass flow. Process data reconciliation based on VDI 2048 is a mathematical-statistic process that makes use of redundant process information. The uncertainty of reconciled feed water flow rates and the thermal reactor output calculated on this basis can be reduced to 0.4 %. The overall process monitored continuously in this manner therefore provides hourly process information of a quality equal to that obtained with acceptance measurements. In the NPP Beznau both methods have been used in parallel to determine the feed water flow rates in 2004 (unit 1) and 2005 (unit 2). Comparison of the results shows that a high level of agreement is obtained between the results of the reconciliation and the results of the tracer measurements. For this reason it was decided that no future tracer measurements will be conducted anymore. A a result of the findings of this comparison, a high level of acceptance of process data reconciliation based on VDI 2048 was achieved.

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The large quantities of information produced within plant processes nearly make the plausibility of data impossible without the help of additional tools. For this reason, a variety of plant monitoring tools has been developed in the past which promise a sensible compression of data. The main problem with the offered tools lies with the omission of procedural plausibility. The newly developed plant monitoring system BTB ProcessPlus is based on the VDI 2048 methodology of process data reconciliation. Plausibility and quality control therefore serve as a basis for the system. With this procedural process image, significant diagnosis and monitoring tools have been developed and now offer a fast and economically optimal support in process optimization. This paper describes the methodology according to VDI 2048. The benefits of the online plant monitoring system are demonstrated by means of examples from day-to-day operations.

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A new neutron shielding resin has been developed for a dual-purpose metal cask. The resin is composed of a cycloaliphatic epoxy, anhydrous acid, catalyst, aluminum hydroxide and boron tetracarbide. Its long-term stability was verified by thermal degradation tests. Estimated weight loss of the resin during storage was about 1 - 2 %. Because the curing reaction of epoxy and curing reagents was moderate at room temperature, a large amount of resin could be treated at one time which would lower fabrication cost. The fabrication process was verified by a full-scale mock-up test. No significant voids or cracks were found in the resin and uniform elemental composition was confirmed.

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Monitoring of VVER fuel assemblies condition in Slovakia is presented in the paper. The Slovak wet interim spent fuel storage facility in NPP Jaslovske Bohunice was build and put in operation in 1986. Since 1999, leak tests of VVER-440 fuel assemblies are provided by special leak tightness detection system "Sipping in Pool" delivered by Framatome-anp facility with external heating for the precise detection of active specimens. Although this system seems to be very effective, the detection time of all fuel assemblies in one storage pool is too long (several months). Therefore, a new "on-line" detection system, based on new sorbent NIFSIL for effective ^<134>Cs and ^<137>Cs activity was developed. The design of this detection system and its possible application in the Slovak wet spent fuel storage facility is discussed in this paper. For completeness, the initial results of the new system are also presented.

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Blowdown process of helium in HCS of CH-HCSB is theoretically analyzed. The influenced area is divided into three volumes and uniform assumption is adopted for VV and main loop of HCS. Analytical correlations of mass and pressure in VV and HCS are obtained. Validation of the analytical model is done by comparison with the numerical results of 3D CFD code. The effect of flow friction is also analyzed. It is found that the flow friction plays important role in the blowdown process and cannot be omitted. The blowdown process in TBM is simulated and analyzed. The results show that the flow in HCS keeps a high value in 20 seconds after break, which will benefit to the removal of decay heat. The results also show that the shock in vacuum vessel caused by pressure rising is not significant.

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Spacer grid springs support the fuel rods in a nuclear fuel system. The spacer grid is a part of the fuel assembly. Since the spring has repeated contact with the fuel rod, fretting wear occurs on the surface of the spring. Design is usually performed to reduce the wear. The conceptual design process for the spring is defined by using the Independence Axiom of axiomatic design and the design is carried out based on the direction that the design matrix indicates. For a detailed design an optimization problem is formulated. In optimization, homologous design is employed to reduce the fretting wear. The deformation of a structure is called homologous if a given geometrical relationship holds for a certain number of structural points before, during, and after the deformation. In this case, the deformed shape of the spring should be the same as that of the fuel rod. This condition is transformed to a function and considered as a constraint in the optimization process. The fretting wear is expected to be reduced due to the homology constraint. The objective function is minimizing the maximum stress to allow a local plastic deformation. Optimization results show that contact occurs in a wide range. Also, the results are verified by nonlinear finite element analysis.

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An analysis system for the long-term mechanical behavior of barrier materials (MACBECE: Mechanical Analysis system considering Chemical transitions of BEntonite-based and CEment-based materials) was developed in order to improve the reliability of the evaluation of the hydraulic field which is one of the important environmental conditions in the safety assessment of the TRU waste disposal. MACBECE is the system that calculates the deformation of barrier materials using their chemical property changes as inputs, and subsequently calculates their hydraulic conductivity taking both their chemical property changes and deformation into consideration. By using MACBECE, the long-term deformation and the transition of hydraulic field for the round-type disposal cavities were evaluated, assuming some sets of chemical evolution data as input. Based on the analysis result, it is considered that the influence of the long-term deformation of the barrier materials on the nuclide migration is not necessarily significant.

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Helium Turbine is used in High Temperature Reactor-helium Gas Turbine (HTR-GT) system, by which the direct helium circulation between the reactor and turbine generator system will come true. Between helium turbine and generator, there is gearbox device which reduces the turbine rotation speed to normal speed of the generator. Three optional gearbox schemes are discussed. (1) Single reduction cylindrical gearbox, which consists of one high speed gear and one low speed gear. Its advantage is simple structure, easy to manufacture, and high reliability, while disadvantage is large volume and misalignment of input and output axle. (2) Planetary gear mechanism with static planet carrier. (3) Planetary gear mechanism with static internal gear. The latter two gearbox devices have similar structure. Their advantage is small volume and high reduction gear ratio, while disadvantage are complicated structure, many gears, low reliability and low mechanical efficiency.

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We have carried out the critical experiments for the MOX fueled tight lattice LWR cores using FCA facility and constructed the XXII-1 series cores. Utilizing the critical experiments carried out at FCA, we have evaluated the reduction of prediction uncertainty in the coolant void reactivity worth of the breeding LWR core based on the bias factor method with focusing on the prediction uncertainty due to cross section errors. In the present study, we have introduced a concept of a virtual experimental value into the conventional bias factor method to overcome a problem caused by the conventional bias factor method in which the prediction uncertainty increases in the case that the experimental core has the opposite reactivity worth and the consequent opposite sensitivity coefficients to the real core. To extend the applicability of the bias factor method, we have adopted an exponentiated experimental value as the virtual experimental value and formulated the prediction uncertainty reduction by the use of the bias factor method extended by the concept of the virtual experimental value. From the numerical evaluation, it has been shown that the prediction uncertainty due to cross section errors has been reduced by the use of the concept of the virtual experimental value. It is concluded that the introduction of virtual experimental value can effectively utilize experimental data and extend applicability of the bias factor method.

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Modular High Temperature Gas-Cooled Reactor (MHTGR) is characterized by inherent safety and higher electrical efficiency, so it can effectively improve the safety and economics of the nuclear power plants. Based upon these advantages, the High Temperature Gas-Cooled Reactor-Pebble Bed Module (HTR-PM) is under design and will be constructed in China to demonstrate the safety and economics of MHTGR. The automatic control system is important and necessary to the safe, economical, and efficient operation of the MHTGR. This paper investigates the control characteristics of the HTGR nuclear power plants, and analyzes the control techniques and existing control strategies of HTGR plants. Advanced control technology which applies modern and intelligent control theory in industrial process provides an opportunity to improve the control performance of the MHTGR plant. Based upon the advanced control technology, the paper proposes a preliminary design concept of hierarchical coordinated control system to the control system design of the HTR-PM which employs the Distributed Control System (DCS) principle.

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Radionuclide migration experiments in quarried blocks of granite under in-situ conditions at the 240-m level in AECL's Underground Research Laboratory (URL) were performed under a five-year cooperative research program between Japan Atomic Energy Research Institute (JAERI, reorganized to Japan Atomic Energy Agency, JAEA) and Atomic Energy of Canada Ltd. (AECL). Migration experiments with Br, ^3H, ^<85>Sr, ^<237>Np, ^<238>Pu, ^<95m>Tc and synthetic colloids, and post-experimental alpha and gamma scanning of the fracture surfaces were performed using 1m^3 granite blocks, containing a single fracture, excavated from a water-bearing fracture zone. The transport of the radionuclides was affected by macroscopic mechanical dispersion, matrix diffusion and element-specific sorption on fracture surfaces. Colloid transport exhibited a complicated process that may include sedimentation and diffusion into stagnant zones.

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We have been studying a dry surface decontamination technology using a low-pressure arc, characterized by features such as less secondary waste. In this report, we investigated the decontamination performance of a low-pressure arc for pipe wastes. A pipe-shaped test piece with corrosion product film containing Co-60 on its inner surface was selected. As a result, the Co-60 removal ratio for the pipe test piece was found to be more than 90 percent. Thus we clarified the possibility of applying low-pressure arc decontamination technology to pipe wastes.

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Experiments on the plastic deformation of 34-mm (nominal) U-shaped pipes and their rupture have been performed in order to extend experiments on the mechanical response of pipe structures to the internal radiolytic gas detonation loads at BWR relevant conditions. Stoichiometric hydrogen-oxygen mixtures were used as a worst case of radiolytic gas composition. Austenitic stainless steel (Werkstoff Nr. 1.4435) and ferritic carbon steel (STPT 410) were used for tested tubes. The dynamic tube response was investigated under radiolytic gas detonation loads at initial pressures up to 70 bar and at room temperature. A dynamic stress-strain diagram from elastic to plastic deformations at strain rates from 100 to 2000 s^<-1> with subsequent rupture of the tested tubes was experimentally obtained, using direct strain measurements and high speed movies. Breaking elongations of 45 - 54% for austenitic 2-mm wall thickness tube were achieved due to radiolytic gas detonation loads at 50 and 57 bar of initial pressure. Ferritic tubes (4.5 mm nominal wall thickness) survived detonations of radiolytic gas at initial pressures of 40 and 70 bar. The maximum deformation achieved was less than 0.4% with a strain amplification factor of 1.2-2.5. The present experimental data are required to study the tube integrity under radiolytic gas detonation loads and to obtain experimental data on the dynamic strain-stress relations for computer code validation.

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As the nuclear renaissance gains momentum, many countries that currently have no nuclear power plants will begin to consider introducing them. It is anticipated that smaller reactors such as the Pebble Bed Modular Reactor (PBMR) will not only be sold to current nuclear states to also to states where there is currently no nuclear experience. A range of issues would have to be considered for nuclear plants to be sold to non-nuclear states, such as the appropriate regulatory environment, standardization and codes, non-proliferation, security of supply, obtaining experienced merchant operators, appropriate financial structures and education and training. The paper considers nine major issues that need to be addressed by governments and vendors alike. International cooperation by organisations such as the IAEA, financial institutions and international suppliers will be required to ensure that developing countries as well as developed ones share the benefits of the nuclear renaissance. The opportunities that the nuclear industry affords to develop local skills, create job opportunities and to develop local manufacturing industries are among the important reasons that the South African Government has decided to support and fund the development of the Pebble Bed Modular Reactor project. These considerations are included in the paper.

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The United States current has a fleet of approximately 100 operating LWRs (thermal) supplying about 20% of the electricity demand of the country. Many of these units have or are anticipated, under plant life extension, to be in service a total of some 60 years. At the same time, forecasts on energy demand and deployment strategies of nuclear power plants (NPPs) over the next 100+ years, indicate that all 100 existing NPPs need to be replaced and as many as 100 additional NPPs need to be added to the fleet in order to meet energy demands and stem further accumulation of GHGs from fossil-fuel based energy sources. Conservatively, at least 20 and more likely 50-75 NPPs need to be newly deployed. Further, to make strides toward energy security and sustainability, the Global Nuclear Energy Project (GNEP) is promoting gradual integration of MOX burning LWRs, fuel reprocessing and deployment of fast reactors (FBRs/FRs) into the global fleet. Consideration of MOX, reprocessing and FBRs merit some consideration with respect to uranium ore resources over the time period of interest. In the present study we developed a basic temporal model over the next 30 to 60 years of existing NPPs and new NPPs to be deployed. Initially, we considered four reactor "categories" over the time period of interest as follows: 1) LWRs (currently operating but) to be decommissioned during the time period, 2) LWRs to be burn MOX fuel during the its operating lifetime, 3) new LWRs to replace decommissioned units and/or added to the fleet and 4) new FRs to replace decommissioned units and/or added to the fleet. To develop the model, we initially used the visual dynamic modeling software, VENSIM. In the study we considered a number of variables such as new NPP (LWRs and FRs) construction rate, decommissioning rate, conversion rate to MOX fueled and overall growth rate of the fleet. We will present results to date on various decommissioning, deployment and electricity demand scenarios relative to forecasts by recent study on the future of nuclear energy.

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In order to investigate the stability of a nuclear reactor core with an oxide mixture of uranium and plutonium (MOX) fuel installed, channel and regional stability tests were conducted with the SIRIUS-F facility. The SIRIUSF facility was designed and constructed to provide a highly accurate simulation of thermal-hydraulic (channel) instabilities and coupled thermalhydraulics-neutronics instabilities of the Advanced Boiling Water Reactors (ABWRs). A realtime simulation was performed by modal point kinetics of reactor neutronics and fuel-rod thermal conduction on the basis of a measured void fraction in a reactor core section of the facility. A time series analysis was performed to calculate decay ratio and resonance frequency from a dominant pole of a transfer function by applying AR methods to the time-series of the core inlet flow rate. Experiments were conducted with the SIRIUS-F facility, which simulates ABWR with MOX fuel installed. The variations in the decay ratio and resonance frequency among the five common AR methods are within 0.03 and 0.01 Hz, respectively. In this system, the appropriate decay ratio and resonance frequency can be estimated on the basis of the Yule-Walker method with the model order of 30.

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An analysis to the hydrogen behavior in the Qinshan Phase II Unit 1 containment was performed by using GASFLOW based on a realistic three dimension containment model. Three cases were taken into consideration, i.e. one case without recombiners and two cases with passive automatic recombiners (PARs) of different arrangement. Results from these three cases were compared, in order to study the effects of the recombiners and their arrangement on the hydrogen distribution and transportation. The difference of the flow field in all the cases was discussed. The characteristics of the hydrogen distribution in different cases were analyzed and the influence of recombiners was presented. The recombiner efficiency in two different arrangements was investigated and discussed.

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In this paper, a mathematical model has been developed by simplifying assumptions typical for natural convective heat transfer from a downward-facing horizontal round plate in confined space. By discretizing the differential equations and setting up the relationship between temporal and spatial nodes, the solution of the resulting differential equation is presented in the form of a correlation between the dimensionless Nusselt and Rayleigh numbers. The numerical result is compared with the experimental measurements of heat transfer from a heated round horizontal plate of 0.3m and 0.1m in diameter performed in water. The numerical and experimental results are in excellent agreement.

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Turbulent flow and temperature fields were determined numerically in a rectangular duct containing a heated rod. As the spacing δ between the rod and the duct wall decreased from 0.10D (D is the rod diameter) to 0.03D, coherent turbulent kinetic energy and temperature fluctuations dramatically increased in the gap region, but, for δ = 0.01D, coherent fluctuations essentially disappeared. As δ/D → 0, the frequency of coherent fluctuations decreased and cross-gap mixing weakened, contrary to predictions based on extrapolated available empirical correlations.

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A vapor explosion can occur by the contact of a hot molten metal liquid with a low boiling temperature cold liquid. The fuel and coolant interaction (FCI) is also postulated to activate the vapor explosion at the severe accidents of a fission nuclear reactor. In order to investigate the mechanism of the vapor explosion, in this study, an observable experiment equipment facility for low-temperature molten materials to be dropped into water was designed. The experiment results show that the molten material temperature has an important effect on the vapor explosion behavior and pressure. The increase of the coolant temperature would decrease the pressure of the vapor explosion.

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